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Journal Articles

Characteristics of global energy confinement in KSTAR L- and H-mode plasmas

Kim, H.-S.*; Jeon, Y. M.*; Na, Y.-S.*; Ghim, Y.-C.*; Ahn, J.-W.*; Yoon, S. W.*; Bak, J. G.*; Bae, Y. S.*; Kim, J. S.*; Joung, M.*; et al.

Nuclear Fusion, 54(8), p.083012_1 - 083012_11, 2014/08

 Times Cited Count:4 Percentile:24.21(Physics, Fluids & Plasmas)

We evaluate the characteristics of global energy confinement in KSTAR ($$tau_{E, rm KSTAR}$$) quantitatively by comparing it with multi-machine scalings, by deriving multiple regression equations for the L- and the H-mode plasmas, and evaluating confinement enhancement of the H-mode phase compared with the L-mode phase in each discharge. From the KSTAR database, $$tau_{E, rm KSTAR}$$ of L-mode plasmas exhibits $$sim 0.04$$ s to $$sim 0.16$$ s and $$tau_{E, rm KSTAR}$$ of H-mode plasmas $$sim 0.06$$ s to $$sim 0.19$$ s. The multiple regression equations derived by statistical analysis present the similar dependency on PL and slightly higher dependency on IP compared with the multi-machine scalings, however the dependency on elongation $$kappa$$ in both L- and H-mode plasmas draw the negative power dependency of $$kappa^{-0.68}$$ and $$kappa^{-0.76}$$ for H-mode and for L- mode database, respectively on the contrary to the positive dependency in all multi-machine empirical scalings. Although the reason is not clear yet, two possibilities are addressed. One is that the wall condition of KSTAR was not clean enough. The other is that striking points on the divertor plate were uncontrolled. For these reasons, as $$kappa$$ increases, the impurities from the wall can penetrate into plasmas easily. As a consequence, the confinement is degraded on the contrary to the expectation of multi-machine scalings.

Journal Articles

Development of high voltage power supply for the KSTAR 170 GHz ECH and CD system

Jeong, J. H.*; Bae, Y. S.*; Joung, M.*; Kim, H. J.*; Park, S. I.*; Han, W. S.*; Kim, J. S.*; Yang, H. L.*; Kwak, J. G.*; Sakamoto, Keishi; et al.

Fusion Engineering and Design, 88(5), p.380 - 387, 2013/06

 Times Cited Count:1 Percentile:12.53(Nuclear Science & Technology)

Journal Articles

Characteristics of the first H-mode discharges in KSTAR

Yoon, S. W.*; Ahn, J.-W.*; Jeon, Y. M.*; Suzuki, Takahiro; Hahn, S. H.*; Ko, W. H.*; Lee, K. D.*; Chung, J. I.*; Nam, Y. U.*; Kim, J.*; et al.

Nuclear Fusion, 51(11), p.113009_1 - 113009_9, 2011/11

 Times Cited Count:25 Percentile:75.77(Physics, Fluids & Plasmas)

Typical ELMy H-mode discharges have been achieved on the KSTAR tokamak with the combined auxiliary heating of NBI and ECRH. The minimum external heating power required is about 1.1 MW at a line-averaged density higher than 1.4$$times$$10$$^{19}$$ m$$^{-3}$$ and a toroidal field of 2 T. There is a clear indication of the increase of the L-H threshold power at densities lower than $$1.4times 10^{19} {rm m}^{-3}$$. The initial analysis of energy confinement time ($$tau$$$$_{E}$$) predicted that $$tau$$$$_{E}$$ was higher than the prediction of multi-machine scaling laws by a factor 1.4-1.6. However, when the contribution of fast ion confinement to the total energy was taken into account, $$tau$$$$_{E}$$ better agreed with the scaling results. A clear increase of electron and ion temperature in the pedestal was observed in the H-mode phase but the core ion temperature did not change significantly. On the other hand, the toroidal rotation also increased over all radii in the H-mode phase. The measured ELM frequency was from 30-50 Hz and the drop of total energy appeared to be less than 5%. Between large ELM spikes, small/grassy ELMs were also identified when mixed heating of NBI and ECRH was applied.

Journal Articles

On maximizing the ICRF antenna loading for ITER plasmas

Mayoral, M.-L.*; Bobkov, V.*; Colas, L.*; Goniche, M.*; Hosea, J.*; Kwak, J. G.*; Pinsker, R.*; Moriyama, Shinichi; Wukitch, S.*; Baity, F. W.*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 11 Pages, 2011/03

For any given ICRF antenna design for ITER, the maximum achievable power strongly depends on the density profiles in the SOL. It has been suggested that gas injection can be used to modify the SOL profiles and thus minimize the sensitivity of the ICRF coupling to variations in the density at the edge of the confined plasma. Recently joint experiments coordinated by the ITPA were performed to characterize further this method. An increase in SOL density during gas injection led to improved coupling for all tokamaks in this multi-machine comparison. The effectiveness of using gas injection over a wide range of conditions, as a tool to tailor the edge density in front of the ICRF antennas, is documented for different gas inlet location and plasma configurations. In addition, any deleterious effects on the confinement and interaction with the antenna near-field are not investigated.

Journal Articles

Commissioning results of the KSTAR neutral beam system

Bae, Y. S.*; Park, Y. M.*; Kim, J. S.*; Han, W. S.*; Kwak, S. W.*; Chang, Y. B.*; Park, H. T.*; Song, N. H.*; Chang, D. H.*; Jeong, S. H.*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 9 Pages, 2011/03

The neutral beam injection (NBI) system is designed to provide the ion heating and current drive for the high performance operation and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR). The KSTAR NBI consists of two beam lines. Each beam line contains three ion sources of which one ion source has been designed to deliver more than 2.5 MW of deuterium neutral beam power with maximum 120-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 14 MW of deuterium beam power with the two beam lines. According to the planned NBI system, the first NBI system is to demonstrate the beam injection from one ion source into the KSTAR tokamak plasma in 2010 campaign including the system commissioning of each components and subsystems. In this paper, the construction and the commissioning of the first NBI system with one ion source is presented.

Journal Articles

Status and result of the KSTAR upgrade for the 2010's campaign

Yang, H. L.*; Kim, Y. S.*; Park, Y. M.*; Bae, Y. S.*; Kim, H. K.*; Kim, K. M.*; Lee, K. S.*; Kim, H. T.*; Bang, E. N.*; Joung, M.*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03

Because the 2010 operation of Korea Superconducting Tokamak Advanced Research (KSTAR) mainly aims to achieve strongly elongated and diverted plasma, all the necessary hardware systems to provide an essential circumstance for the plasma shaping were newly installed and upgraded in 2010. In this paper, general configuration of the upgraded systems described earlier will be outlined. Moreover, several key performances and test results of the systems will be also reported in summary.

Oral presentation

Confinement characteristics of the extended operation regime of KSTAR toward advanced scenarios

Na, Y. S.*; Suzuki, Takahiro; Ide, Shunsuke; Mueller, D.*; Kim, J. H.*; Miyata, Yoshiaki; Kim, S. H.*; Kim, H. S.*; Jeon, Y. M.*; Bae, Y. S.*; et al.

no journal, , 

Development of advanced scenarios, an important experimental goal for the KSTAR project, has just begun. Target plasmas were successfully produced using large bore plasma and early divertor formation which exhibit low internal inductance with low magnetic shear at the centre and no sawtooth instability. Auxilliary heating during the current rampup phase was employed to slow the inductive current diffusion to the centre of the plasma. With respect to hybrid scenario development, so-called "Ip-overshoot" method being used in JET is applied for tailoring magnetic shear at reduced plasma current for higher poloidal beta and bootstrap current fraction. The confinement characteristics of these scenarios are investigated. Transport modeling is performed self-consistently with an integrated simulation package incorporating plasma equilibrium, transport, heating and current drive. Firstly, the current rampup phase is simulated and its impact on the target q-profile is addressed. Secondly, energy confinement of flattop phases is discussed. In addition, the non-inductive current drive fraction including the bootstrap current fraction is calculated. Lastly, these scenarios are compared with advanced scenarios developed in other tokamak devices and future directions in achieving advanced regimes are discussed.

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